The lithium blanket surrounding the core of neutron generating devices such as fusion and advanced fission reactors can advantageously be utilized for both heat transfer and production of tritium. Tritium already proves useful in applications such as lighting and weapons production, and is expected to be the primary fuel source for fusion reactors in deuterium/tritium (D/T) fueled power plants.
The formation of tritium occurs upon exposure of the molten lithium metal of the blanket to the neutron flux of a conventional fission reactor utilizing, for instance, U233, U235, or Pu239, upon which exposure lithium tritide (LiT) will form in the blanket with the tritium bred by neutron reaction with lithium atoms. A single lithium-6 atom exposed to the thermal neutron flux can fission to produce a tritium nucleus and a helium nucleus. Lithium-7 can likewise be used as it will fission on capturing a fast neutron to produce a tritium nucleus, an □-particle and a neutron, though lithium-7 has a considerably smaller cross section for low-energy neutron flux than
Ideally, since tritium is unavailable in any significant quantities in nature, a practical and efficient D/T fueled power plant will be developed in which the production and recovery of tritium from the lithium blanket can be carried out at least at the same rate as it is consumed as fuel in the energy production process.
It is generally desirable that only small amounts of tritium be allowed to build up within the lithium blanket before the recovery rate matches the breeding rate. If the excess tritium is not removed from the blanket, the rate of tritium permeation through the blanket and heat exchanger structures can increase, which can pose a radioactivity problem. Large quantities of tritium in the blanket system can also increase the radioactivity hazards during routine maintenance and emergencies associated with mechanical and structural failures. As such, methods for separation and recovery of tritium from the lithium blanket are necessary.
Various processes have been evaluated tor recovering tritium from the lithium blanket and/or coolant systems within fusion reactors. Methods have included gettering, permeation, fractional distillation, cold trapping, and liquid extraction followed by electrolysis. The latter has been the most popular for development, in which a molten salt extraction of the LiT is carried out followed by electrolysis in which the molten salt is utilized as the electrolyte leading to reduction of the lithium ion to form lithium metal at one electrode and oxidation of the tritide ion at the other electrode to form tritium.
Unfortunately, current extraction/electrolysis processes require high speed and high temperature centrifugal separators, which are expensive and complicated to operate. In addition, the materials involved in the extraction process can be highly corrosive, which adds costs as well as safety issues. Moreover, salts used in the liquid/liquid extraction can solubilize into the blanket solution, which can have an impact on tritium breeding due to neutron interaction with the salts.
In view of such issues, what are needed in the art are methods and systems for safe and effective recovery of tritium from the molten lithium metal blanket that can prove to be more economical and straight forward as compared to previously known methods so as to reduce both capital costs and operating expenses.